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Notice

Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

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Information about this document as published in the Federal Register.

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Background

Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission (the Commission or NRC staff) is publishing this regular biweekly notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act. This provision grants the Commission the authority to issue and make immediately effective any amendment to an operating license upon a determination by the Commission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Commission of a request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or Start Printed Page 46240proposed to be issued from, July 11, 2003, through July 24, 2003. The last biweekly notice was published on July 22, 2003 (68 FR 43382).

Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards; Consideration Determination, and Opportunity for a Hearing

The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendment would not (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination.

Normally, the Commission will not issue the amendment until the expiration of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Commission may issue the license amendment before the expiration of the 30-day notice period, provided that its final determination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the Commission take this action, it will publish in the Federal Register a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and should cite the publication date and page number of this Federal Register notice. Written comments may also be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of written comments received may be examined at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The filing of requests for a hearing and petitions for leave to intervene is discussed below.

By September 4, 2003, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written request for a hearing and a petition for leave to intervene. Requests for a hearing and a petition for leave to intervene shall be filed in accordance with the Commission's “Rules of Practice for Domestic Licensing Proceedings” in 10 CFR part 2. Interested persons should consult a current copy of 10 CFR 2.714, which is available at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​doc-collections/​cfr/​. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition; and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) The nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect(s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to 15 days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter. Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the petitioner shall provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert opinion which support the contention and on which the petitioner intends to rely in proving the contention at the hearing. The petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the petitioner intends to rely to establish those facts or expert opinion. Petitioner must provide sufficient information to show that a genuine dispute exists with the applicant on a material issue of law or fact. Contentions shall be limited to matters within the scope of the amendment under consideration. The contention must be one which, if proven, would entitle the petitioner to relief. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves no significant hazards consideration, the Commission may issue the amendment and make it immediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.

If the final determination is that the amendment request involves a significant hazards consideration, any Start Printed Page 46241hearing held would take place before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff, or may be delivered to the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, by the above date. Because of continuing disruptions in delivery of mail to United States Government offices, it is requested that petitions for leave to intervene and requests for hearing be transmitted to the Secretary of the Commission either by means of facsimile transmission to 301-415-1101 or by e-mail to hearingdocket@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and because of continuing disruptions in delivery of mail to United States Government offices, it is requested that copies be transmitted either by means of facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov. A copy of the request for hearing and petition for leave to intervene should also be sent to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for a hearing will not be entertained absent a determination by the Commission, the presiding officer or the Atomic Safety and Licensing Board that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the application for amendment which is available for public inspection at the Commission's PDR, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management System's (ADAMS) Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

Carolina Power & Light Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina

Date of amendments request: July 21, 2003.

Description of amendments request: The proposed license amendment requests approval to revise the Updated Final Safety Analysis Report, Section 9.4.5, “Turbine Building Ventilation System,” and supporting information in Section 6.4.4.1, “Radiological Protection,” and Section 15.6.3, “Main Steam Line Break Accident,” to allow the system to be operated in a once-through versus recirculation configuration in support of outage activities.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The accident of concern for the proposed modification is a Main Steam Line Break (MSLB). The probability of this event is not impacted by the change to the turbine building ventilation system configuration. The consequences of the event have been re-evaluated to determine the impact on control room operator doses and offsite doses. The re-evaluation was performed consistent with the analysis done in support of the adoption of Alternative Source Term (AST) which was approved for use at BSEP [Brunswick Steam Electric Plant] in Amendments 221 and 246 for Units 1 and 2, respectively. The results of the re-evaluation demonstrate that control room doses remain well below regulatory limits and [that] there is no significant impact on offsite doses. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The turbine building ventilation system is non-safety related and its purpose is to provide an acceptable environment for equipment and personnel within the turbine building as well as treat the gaseous effluent prior to release. As such, modification of this system cannot (1) Alter any design basis accident initiators, (2) create new types of accident precursors, or (3) introduce new failure modes of safety related equipment. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The margin of safety for this modification is considered to be that provided by meeting the applicable regulatory limits. Operation of the turbine building ventilation system in a once-through versus recirculation configuration does not impact the ability to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room, remain well within corresponding regulatory limits with respect to a MSLB event (i.e., the only event whose consequences can be impacted by the proposed modification). This was confirmed through re-evaluation of the consequences of a MSLB event, consistent with the analysis done in support of the adoption of AST. Since the proposed changes continue to ensure that the doses at the exclusion area and low population zone boundaries, as well as the control room are within corresponding regulatory limits, the proposed license amendment does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Steven R. Carr, Associate General Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Allen G. Howe.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: April 25, 2003, as supplemented on May 21 and June 11, 2003.

Description of amendment request: The proposed changes to the Technical Specifications (TSs) consists of revisions to protective instrumentation specifications. These changes are made to resolve non-conservative TS issues, relax overly restrictive requirements, and to provide consistency between TS and design and licensing bases. These changes also involve reformatting data, as well as relocation of some data to plant-controlled documents.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The NRC staff's review is presented below:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. Start Printed Page 46242

The proposed changes do not significantly affect the design or fundamental operation and maintenance of the plant. Accident initiators or the frequency of analyzed accident events are not significantly affected as a result of the proposed changes; therefore, there will be no significant change to the probabilities of accidents previously evaluated.

The proposed changes do not significantly alter assumptions or initial conditions relative to the mitigation of an accident previously evaluated. The proposed changes continue to ensure process variables, structures, systems, and components (SSCs) are maintained consistent with the safety analyses and licensing basis. The revised TSs continue to require that SSCs are properly maintained to ensure operability and performance of safety functions as assumed in the safety analyses. The design basis events analyzed in the safety analyses will not change significantly as a result of the proposed changes to the TSs.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not involve any physical alteration of the plant (no new or different type of equipment being installed) and do not involve a significant change in the design, normal configuration or basic operation of the plant. The proposed changes do not introduce any new accident initiators. In some cases, the proposed changes impose different requirements; however, these new requirements are consistent with the assumptions in the safety analyses and current licensing basis. Where requirements are relocated to other licensee-controlled documents, adequate controls exist to ensure their proper maintenance.

Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Involve a significant reduction in a margin of safety.

Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following an accident situation. The proposed changes do not significantly affect any of the assumptions, initial conditions or inputs to the safety analyses. Plant design is unaffected by these proposed changes and will continue to provide adequate defense-in-depth and diversity of safety functions as assumed in the safety analyses.

There is no proposed change to Safety Limits and only administrative and more restrictive changes to Limiting Safety System Setting requirements. The proposed changes maintain requirements consistent with safety analyses assumptions and the licensing basis. Fission product barriers will continue to meet their design capabilities without significant impact to their ability to maintain parameters within acceptable limits. The safety functions are maintained within acceptable limits without any significant decrease in margin.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037-1128.

NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Date of amendment request: February 27, 2003. By a letter dated July 17, 2003, the licensee revised its analysis about the issue of no significant hazards consideration.

Description of amendment request: The proposed amendments would revise the Technical Specifications (TS) Section 3.4.9, Reactor Coolant System Pressure and Temperature (P/T) Limits, and delete the license conditions specified in Facility Operating License Sections 2.C(8) and 3.P, Pressure-Temperature Limit Curves, for Dresden Nuclear Power Station, Units 2 and 3 respectively. The P/T limit curves are proposed to be replaced with ones that are applicable to the remainder of the licensed life of the plant.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. The proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The P/T limits are prescribed during all operational conditions to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate, resulting in non-ductile failure of the reactor coolant pressure boundary, which is an unanalyzed condition. The methodology used to determine the P/T limits has been approved by the NRC and thus is an acceptable method for determining these limits. Therefore, the proposed changes do not affect the probability of an accident previously evaluated.

There is no specific accident that postulates a non-ductile failure of the reactor coolant pressure boundary. The loss of coolant accident analyzed for the plant assumes a complete break of the reactor coolant pressure boundary. The revision to the P/T limits does not change this assumption. Thus, the radiological consequences of any accident previously evaluated are not increased.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not change the response of plant equipment to transient conditions. The proposed changes do not introduce any new equipment, modes of system operation, or failure mechanisms.

Non-ductile failure of the reactor coolant pressure boundary is not an analyzed accident. The proposed changes to the P/T limits were developed using an NRC-approved methodology, and thus the revised limits will continue to provide protection against non-ductile failure of the reactor coolant pressure boundary.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. The proposed TS changes do not involve a significant reduction in a margin of safety.

The margin of safety related to the proposed changes is the margin between the proposed P/T limits and the pressures and temperatures that would produce non-ductile failure of the reactor coolant pressure boundary. The use of an NRC-approved methodology together with conservatively-chosen plant-specific input parameters provides an acceptable margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the requested amendments involve no significant hazards consideration.

Attorney for licensee: Mr. Edward J. Cullen, Deputy General Counsel, Exelon BSC—Legal, 2301 Market Street, Philadelphia, PA 19101.

NRC Section Chief: Anthony J. Mendiola. Start Printed Page 46243

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit 3, York County and Lancaster County, Pennsylvania

Date of amendment request: June 23, 2003.

Description of amendment request: Exelon Generation Company, LLC, the licensee, is proposing a change to the Peach Bottom Atomic Power Station (PBAPS), Unit 3, Technical Specifications (TSs) contained in Appendix A to the Operating License. This proposed change will revise the TS section on safety limits to incorporate revised safety limit minimum critical power ratios (SLMCPRs) based on the cycle-specific analysis performed by Global Nuclear Fuel for PBAPS, Unit 3, Cycle 15.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's analysis against the standards of 10 CFR 50.92(c). The NRC staff's review is presented below.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Changing the SLMCPRs does not require any physical plant modifications, physically affect any plant components, or involve changes in plant operation. Therefore, the probability of an accident previously evaluated remains unchanged.

The operability of plant systems designed to mitigate any consequences of accidents has not changed, therefore, the consequences of an accident previously evaluated are not expected to increase.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve any modifications of the plant configuration for allowable modes of operation. The SLMCPRs are not accident initiators, and their revision will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

The proposed SLMCPRs provide a margin of safety by ensuring that no more than 0.1% of the rods are in a boiling transition if the operating limit minimum critical power ratios are violated during all modes of operation. The change in the SLMCPRs continues to ensure that during normal operation and during abnormal operational transients, at least 99.9% of all fuel rods in the core do not experience transition boiling if the limit is not violated when all uncertainties are considered, thereby preserving the fuel cladding integrity. Therefore, the proposed TS change will not involve a significant reduction in a margin of safety.

Based on the NRC staff's review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Mr. Edward Cullen, Vice President & General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, Kennett Square, PA 19348.

NRC Section Chief: James W. Clifford.

FirstEnergy Nuclear Operating Company, et al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1 (BVPS-1), Beaver County, Pennsylvania

Date of amendment request: June 24, 2003.

Description of amendment request: The proposed amendments would change the Technical Specification (TS) steam generator tube inspection definition such that the definition of tube inspection would exclude the portion of the tube within the tubesheet below the W* distance and would change the tube plugging criteria to indicate that the plugging or repair criteria does not apply to service-induced degradation identified in the W* distance. Service-induced degradation identified in the W* distance would be repaired upon detection. The W* distance is defined in Westinghouse Topical Report, WCAP-14797, Revision 1, and is the distance from the top of the tubesheet to the bottom of the W* length including the distance to the bottom of the WEXTEX transition (approximately 0.25 inches from the top of the tubesheet) plus uncertainties. This equals approximately 7.12 inches on the hot leg side plus the distance to the bottom of the WEXTEX transition and 7.62 inches on the cold leg side plus the distance to the bottom of the WEXTEX transition.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed change modifies the BVPS Unit 1 TSs to incorporate [an] SG [steam generator] tube inspection scope based on WCAP-14797, Revision 1. The proposed change only clarifies the current process which has been utilized in the past. The W* analysis takes into account the reinforcing effect that the tubesheet has on the external surface of an expanded SG tube. Tube-bundle integrity will not be adversely affected by the implementation of the W* tube inspection scope. SG tube burst or collapse cannot occur within the confines of the tubesheet; therefore, the tube burst and collapse criteria of Regulatory Guide (RG) 1.121 are inherently met. Any degradation below the W* distance is shown by analysis and test results to be acceptable, and therefore does not result in an increase in probability of a tube rupture or an increase in the consequences of a tube rupture.

Tube burst is precluded for cracks within the tubesheet by the constraint provided by the tubesheet. However, in the unlikely event of a complete circumferential separation of a tube occurring below the W* distance, SG tube pullout is precluded, tube integrity is maintained and leakage is predicted to be maintained within the Updated Final Safety Analysis Report limits during all plant conditions.

In conclusion, the incorporation of the W* inspection scope into BVPS Unit 1 TS[s] maintains existing design limits and does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed change modifies the BVPS Unit 1 TSs to incorporate SG tube inspection scope based on WCAP-14797, Revision 1. Tube-bundle integrity will be maintained during all plant conditions upon implementation of the proposed tube inspection scope. Use of this scope does not induce a new mechanism that would result in a different kind of accident from those previously analyzed. Even with the limiting circumstances of a complete circumferential separation of a tube occurring below the W* distance, SG tube pullout is precluded and leakage is predicted to be maintained within the design limits during all plant conditions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

No. WCAP-14797, Revision 1 describes the testing that was performed to define the length of non-degraded tubing that is sufficient to compensate for the axial forces on the tube and thus prevent pullout. The operating conditions utilized in WCAP-14797, Revision 1, bound BVPS Unit 1 operating conditions. Upon implementation of the W* inspection scope, operation with potential cracking below the W* distance in Start Printed Page 46244the WESTEX expansion region of the SG tubing meets the margin of safety as defined in RG 1.121 and RG 1.83 and the requirements of General Design Criteria 14, 15, 16, 31, and 32.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for Licensee: Mary O'Reilly, FirstEnergy Nuclear Operating Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH 44308.

NRC Section Chief: Richard J. Laufer.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit 3 Nuclear Generating Plant, Citrus County, Florida.

Date of amendment request: July 14, 2003.

Description of amendment request: The proposed license amendment would revise Technical Specification 3.7.9 by adding a note to allow a one-time 10-day completion time for restoring an inoperable nuclear services seawater system train to operable status. The proposed change would allow the refurbishment of one nuclear services seawater system emergency pump (RWP-2A or RWP-2B) online. The note would specify that the one-time 10-day completion time will expire on December 30, 2004.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed license amendment extends, on a one-time basis, the Completion Time for restoring an inoperable Nuclear Services Seawater System train to Operable status. The Nuclear Services Seawater System is designed to provide cooling for components essential to the mitigation of plant transients and accidents. The system is not an initiator of design basis accidents. During the requested extended time period of ten days, the redundant Emergency Nuclear Services Seawater pump will be available and capable of providing cooling for containment heat loads and essential equipment during emergency conditions. RWP-1 is the CR-3 [Crystal River Unit 3] normal duty Nuclear Closed Cycle Cooling Water pump. Although RWP-1 is non-safety related and its motor is non-seismic, has a lower flow capability than either RWP-2B or RWP-2A and is not connected to an emergency power source, it will also be available and capable of removing emergency heat loads from essential equipment from all design basis events. Informal calculations performed show that below a Ultimate Heat Sink (UHS) temperature of approximately 90°F, RWP-1 can maintain adequate heat removal under accident conditions.

A Probabilistic Safety Assessment (PSA) has been performed to assess the risk impact of an increase in Completion Time. Although the proposed one-time change results in an increase in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), the value of these increases are considered as very small in the current regulatory guidance.

Therefore, granting this LAR [License Amendment Request] does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does not create the possibility of a new or different type of accident from any accident previously evaluated.

The proposed license amendment extends, on a one-time basis, the Completion Time for restoring an inoperable Nuclear Services Seawater System train to Operable status.

The proposed LAR will not result in changes to the design, physical configuration of the plant or the assumptions made in the safety analysis. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does not involve a significant reduction in the margin of safety.

The proposed license amendment extends, on a one-time basis, the Completion Time for restoring an inoperable Nuclear Services Seawater System train to Operable status. The proposed change will allow online repair of one of the Emergency Nuclear Services Seawater pumps to improve its reliability and useful lifetime, thus increasing the long term margin of safety of the system.

The proposed LAR will reduce the probability (and associated risk) of a plant shutdown to repair an Emergency Nuclear Services Seawater pump. To ensure defense in depth capabilities and the assumptions in the risk assessment are maintained during the proposed one-time extended Completion Time, CR-3 will continue the performance of 10 CFR 50.65(a)(4) assessments before performing maintenance or surveillance activities and no maintenance activities of other risk sensitive equipment beyond that required for the refurbishment activity will be scheduled concurrent with the repair activity. Other compensatory actions that may be implemented, include: Use of pre-job briefings and periodic operator walkdowns to assess status of risk sensitive equipment in the redundant train, selection of beneficial Makeup Pump configurations and redundant off-site power feeds to the remaining Emergency Nuclear Services Seawater System pump, no elective maintenance to be scheduled in the switchyard, and the establishment of fire watches in fire areas identified in [PSA Risk Assessment of RWP-2A/2B Extended AOT [Allowed Outage Time]].

As described above in Item 1, a PSA has been performed to assess the risk impact of an increase in Completion Time. Although the proposed one-time change results in an increase in Core Damage Frequency (CDF), and Large Early Release Frequency, the value of these increases are considered as very small in the current regulatory guidance.

Therefore, granting this LAR does not involve a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Steven R. Carr, Associate General Counsel—Legal Department, Progress Energy Service Company, LLC, Post Office Box 1551, Raleigh, North Carolina 27602.

NRC Section Chief: Allen G. Howe.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear Power Plant, Kewaunee County, Wisconsin.

Date of amendment request: July 7, 2003.

Description of amendment request: The proposed amendment would revise the Kewaunee Nuclear Power Plant Technical Specification (TS) Section 3.3.e, “Service Water System,” to add requirements for the turbine building service water header isolation logic.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration which is presented below:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated.

The service water system and specifically the supply to the turbine building, does not initiate any accidents previously evaluated. This change will provide an automatic feature to a function that was previously available to operators, to ensure Emergency Safety Features (ESF) loads will receive adequate service water flow. Flow is provided to ESF components that are cooled by service water without relying on the operator to identify and take action to provide isolation. Diesel loading and sequencing will not be adversely affected by this change. The components supplied by the service water system will continue to be supplied in a timely manner. The valve logic will be properly calibrated and tested consistent with other valves associated with safety significant structures, systems and components.

Therefore, the proposed change will not increase the probability or consequences of an accident previously evaluated. Start Printed Page 46245

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

This change will not affect the service water system function or any components that are accident initiators. The ability to isolate the turbine building load in the event of a system malfunction has been previously evaluated.

Therefore, any change to the system would not affect the probability of an accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

This change will ensure that Engineered Safety Features (ESF) components receiving service water-cooling are not negatively impacted by turbine building load. There are no components served by the turbine building header that are safety systems, structures, or components.

Therefore, NMC concludes that there is not a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, P.O. Box 1497, Madison, WI 53701-1497.

NRC Section Chief: L. Raghavan.

PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

Date of amendment request: July 1, 2003.

Description of amendment request: The proposed amendment would change the Unit 1 Technical Specifications (TSs) by including the Unit 1 Cycle 14 (U1C14) Minimum Critical Power Ratio (MCPR) Safety Limits in Section 2.1.1.2, changing the references listed in Section 5.6.5.b, and changing the design features in Section 4.2.1.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?

Response: No.

The proposed change to the MCPR Safety Limits does not directly or indirectly affect any plant system, equipment, component, or change the processes used to operate the plant. Further, the U1C14 MCPR Safety Limits are generated using NRC approved methodology and meet the applicable acceptance criteria. Thus, this proposed amendment does not involve a significant increase in the probability of occurrence of an accident previously evaluated.

Prior to the startup of U1C14, licensing analyses are performed (using NRC approved methodology referenced in Technical Specification Section 5.6.5.b) to determine changes in the critical power ratio as a result of anticipated operational occurrences. These results are added to the MCPR Safety Limit values proposed herein to generate the MCPR operating limits in the U1C14 COLR [Core Operating Limits Report]. These limits could be different from those specified for the U1C13 COLR. The COLR operating limits thus assure that the MCPR Safety Limit will not be exceeded during normal operation or anticipated operational occurrences. Postulated accidents are also analyzed prior to startup of U1C14 and the results shown to be within the NRC approved criteria.

The U1C14 reload fuel bundles will utilize a small amount of depleted uranium in certain fuel rods, in addition to natural and slightly enriched uranium. There is no change to the composition of the fuel pellets containing depleted uranium material (i.e., UO2) except a slight decrease in the amount of Uranium-235. Therefore, the use of depleted uranium in the fuel rods does not affect the mechanical performance of the fuel rods. The depleted uranium was modeled in the approved design and licensing methodology.

The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U1C14 core operating limits. The use of this approved methodology does not increase the probability of occurrence or consequences of an accident previously evaluated.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The change to the MCPR Safety Limits does not directly or indirectly affect any plant system, equipment, or component and therefore does not affect the failure modes of any of these items. Thus, the proposed changes do not create the possibility of a previously unevaluated operator error or a new single failure.

The use of depleted uranium in the fuel rods does not affect the mechanical performance of the fuel rods.

The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U1C14 core operating limits. The use of this approved methodology does not create the possibility of a new or different kind of accident.

Therefore, the proposed amendment does not involve the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Since the proposed changes do not alter any plant system, equipment, component, or the processes used to operate the plant, the proposed change will not jeopardize or degrade the function or operation of any plant system or component governed by Technical Specifications. The proposed MCPR Safety Limits do not involve a significant reduction in the margin of safety as currently defined in the Bases of the applicable Technical Specification sections, because the MCPR Safety Limits calculated for U1C14 preserve the required margin of safety.

The use of depleted uranium in the fuel rods does not affect the mechanical performance of the fuel rods.

The changes to the references in Section 5.6.5.b were made to properly reflect the NRC approved methodology used to generate the U1C14 core operating limits. This approved methodology is used to demonstrate that all applicable criteria are met, thus, demonstrating that there is no reduction in the margin of safety.

Therefore, these proposed changes do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, Allentown, PA 18101-1179.

NRC Section Chief: Richard J. Laufer.

PSEG Nuclear, LLC, Docket No. 50-354, Hope Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 9, 2003.

Description of amendment request: The proposed amendment would revise Technical Specification (TS) 3.7.2, to increase the allowed outage time (AOT) for one train of the control room emergency filtration (CREF) system from 7 days to 30 days. The proposed AOT change would only apply when one CREF train is inoperable due to an inoperable chiller during Modes 1, 2, or 3.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (10 CFR) section 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No. Start Printed Page 46246

The proposed TS change does not affect the design, operational characteristics, function or reliability of the control room emergency filtration (CREF) system. The CREF is not an initiator of any previously evaluated accident. The proposed change will increase the allowed outage time for the chiller from seven days to 30 days for the chiller in OPERATIONAL CONDITIONS 1, 2, AND 3. The 30-day AOT is based on the low probability of an event requiring control room isolation concurrent with failure of the redundant train. Therefore, one train will always be available to remove the normal and accident heat loads and provide control room isolation.

Increasing the AOT will allow for completion of maintenance activities requiring extended down time to perform and result in significant improvements to the overall reliability of control room chillers. Improving reliability will provide additional assurance that chillers will be capable of performing their design basis accident function.

Therefore, this proposed amendment does not involve a significant increase in the probability of occurrence or radiological consequences of an accident previously analyzed.

2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change will increase the AOT for the control room chiller from seven to thirty days in modes 1 through 3. During the time one chiller is inoperable, the redundant train is capable of handling the heat loads during normal operation and accident conditions. The proposed change does not involve a change in the design, configuration, or method of operation of the plant that could create the possibility of a new or different kind of accident. The proposed change would not introduce new failure modes or effects and would not, in the absence of other unrelated failures, create a new or different accident from any accidents previously evaluated.

Therefore, the proposed changes would not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the change involve a significant reduction in the margin of safety?

Response: No.

The basis for technical specification 3/4.7.2 is to ensure that the temperature in the control room does not exceed the maximum allowable for the equipment and instrumentation located therein. The system also limits radiation exposure to control room personnel following an accident to below GDC-19 [General Design Criterion 19] limits. Either of the two redundant trains can perform these functions. Although one chiller may be inoperable for longer than seven days, the redundant train can perform all normal and accident functions. The length of time for the chiller AOT is sufficiently short to assure that an event requiring control room isolation concurrent with the failure of the redundant train is not credible.

Therefore, these changes do not involve a significant reduction in [a] margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

Date of amendment request: June 6, 2003.

Description of amendment request: The proposed amendment would revise the Salem Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specification (TS) 3/4.11.2.5, “Explosive Gas Mixture.” The proposed changes would: (1) Add a footnote to Limiting Condition for Operation (LCO) 3.11.2.5, to allow maintenance on the waste gas system; (2) revise Surveillance Requirement 4.11.2.5, to delete reference to hydrogen which is not limited by the LCO; and (3) incorporate changes to the appropriate TS Bases pages.

Basis for proposed no significant hazards consideration determination: As required by Title 10 of the Code of Federal Regulations (10 CFR) section 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to the Technical Specifications (TS) 3/4.11.2.5, Explosive Gas Mixtures, would correct inconsistencies while continuing to preclude the combination of explosive concentrations of oxygen and hydrogen in the Salem Generating Station (SGS) Unit 1 and 2 waste gas system. The changes eliminate the potential for misinterpretation and achieve internal consistency between TS sections. No changes to the design of structures, systems, or components (SSC) are made and there are no effects on accident mitigation.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Section 15.3.6 of the SGS Updated Final Safety Analysis Report (UFSAR) summarizes the results of a postulated non-mechanistic rupture of a waste gas decay tank. This postulated accident scenario is not affected by the proposed amendment, nor is any new accident scenario introduced by the proposed changes. The proposed administrative and editorial changes to the TS do not change the design function of or operation of any SSCs. The TS, as amended, would continue to limit explosive and flammable gas concentrations to prevent an uncontrolled release from the waste gas system.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed changes [ ] do not affect the ability of plant SSCs to perform their design basis accident functions. In addition, the [proposed TS license amendment] does not change the margin of safety since no SSCs are changed and the [current] limits on explosive gas mixtures are maintained.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business Unit—N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Section Chief: James W. Clifford.

TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak Steam Electric Station, Units 1 and 2, Somervell County, Texas.

Date of amendment request: July 10, 2003.

Brief description of amendments: The proposed amendments would provide for a one-time change for each unit to revise Technical Specification 3.7.10, entitled “Control Room Emergency Filtration/Pressurization System (CREFS),” to extend the COMPLETION TIME for ACTION B from 24 hours to 14 days.

Basis for proposed no significant hazards consideration determination: As required by Section 50.91(a) of Title 10 of the Code of Federal Regulations (10 CFR), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated? Start Printed Page 46247

Response: No.

This is a revision to the Technical Specifications for the control room emergency/filtration system which is a mitigation system designed to minimize in leakage and to filter the control room atmosphere to protect the operator following accidents previously analyzed. An important part of the system is the control room boundary. The control room boundary integrity is not an initiator or precursor to any accident previously evaluated. Therefore, the probability of any accident previously evaluated is not increased. The analysis of the consequences of analyzed accident scenarios under the control room breach conditions along with the compensatory actions for restoration of control room integrity demonstrate that the consequences of any accident previously evaluated are not increased. Therefore, it is concluded that this change does not significantly increase the probability of an accident previously evaluated.

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes will not impact the accident analysis. The changes will not alter the requirements of the control room emergency/filtration system or its function during accident conditions. The administrative controls and compensatory actions will ensure the control room emergency/filtration system will perform its safety function. [Sentence deleted] The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions and current plant operating practice. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed changes do not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by these changes. The proposed changes will not result in plant operation in a configuration outside the design basis for an unacceptable period of time without compensatory actions and administrative controls. The proposed changes do not affect systems that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. Therefore, the proposed change does not involve a reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC 20036.

NRC Section Chief: Robert A. Gramm.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

Date of amendment request: September 5, 2002, as supplemented April 16, June 9, and July 7, 2003.

Description of amendment request: The proposed technical specification (TS) amendment will add provisions to permit inspection and related repair of a buried fuel oil storage tank during plant operation.

Basis for proposed no significant hazards consideration determination: As required by 10 CFR 50.91(a), the licensee has provided its analysis of the issue of no significant hazards consideration, which is presented below:

1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

This proposed TS change does not alter the assumptions of the accident analyses or the TS Basis. The inclusion of provisions to permit inspection and related repair of a buried fuel oil storage tank during plant operation does not impact the availability of the EDGs [emergency diesel generators] to perform their required function, which is to provide an emergency source of power to vital equipment when a normal power source is not available. Furthermore, while a buried tank is out of service, the proposed change includes requirements to verify the availability of onsite and offsite fuel oil sources to ensure that an adequate supply of fuel oil remains available. Therefore, the proposed change does not result in a significant increase in either the probability or consequences of an accident previously evaluated.

2. Does the proposed license amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

This proposed TS change does not involve a physical change to the plant, nor does it alter the assumptions of the accident analyses. Inclusion of provisions to permit inspection and related repair of a buried fuel oil storage tank does not introduce any new failure modes. Therefore, the proposed change does not create the possibility of a new or different kind of accident from those previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

This proposed TS change alters the method of operation of the Fuel Oil System. However, the availability of the EDGs to perform their required function is not impacted, and the assumptions of the accident analyses are not altered. Furthermore, a plant specific risk evaluation of the acceptability of the provisions was performed. The risk evaluation concluded that the risk impact is acceptable (i.e., is characterized as “very small” by Regulatory Guide 1.174 criteria and is within the acceptance criteria of Regulatory Guide 1.177). Therefore, the proposed change does not significantly reduce the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to determine that the amendment request involves no significant hazards consideration.

Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, Dominion Resources Services, Inc., Millstone Power Station, Building 475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.

NRC Section Chief: John A. Nakoski.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for A Hearing in connection with these actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) The applications for amendment, (2) the amendment, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment as indicated. All of these items are available for public inspection Start Printed Page 46248at the Commission's Public Document Room, located at One White Flint North, Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible from the Agencywide Documents Access and Management Systems (ADAMS) Public Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/​reading-rm/​adams.html. If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to pdr@nrc.gov.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, Maricopa County, Arizona.

Date of application for amendments: August 28, 2002.

Brief description of amendments: The amendments extend the expiration date of the operating licenses from December 31, 2024, to June 1, 2025, for Unit 1, December 9, 2025, to April 24, 2026, for Unit 2, and March 25, 2027, to November 25, 2027, for Unit 3 of Palo Verde Nuclear Generating Station.

Date of Issuance: July 15, 2003.

Effective date: July 15, 2003, and shall be implemented within 60 days of the date of issuance.

Amendment Nos.: Unit 1-147, Unit 2—147, Unit 3—147.

Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The amendments revised the Operating Licenses.

Date of initial notice in Federal Register: October 15, 2002 (67 FR 63688). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 15, 2003.

No significant hazards consideration comments received: No.

Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (CCNPP), Calvert County, Maryland.

Date of application for amendments: June 11, 2002, as supplemented May 2, 2003, and June 23, 2003.

Brief description of amendments: These amendments revise the CCNPP Technical Specification Administrative Controls Section to incorporate six changes previously approved for the Improved Standard Technical Specifications and one administrative change in renumbering pages.

Date of issuance: July 16, 2003.

Effective date: As of the date of issuance to be implemented within 30 days.

Amendment Nos.: 259 and 236.

Renewed Facility Operating License Nos. DPR-53 and DPR-69: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 3, 2002 (67 FR 56318). The May 9, 2003, letter provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the initial proposed no significant hazards consideration determination as published in the Federal Register on September 3, 2002 (67 FR 56318). The June 23, 2003, letter withdrew the requested change dealing with clarifying references to 10 CFR part 20 in the Technical Specifications and did not change the initial proposed no significant hazards consideration determination as published in the Federal Register on September 3, 2002 (67 FR 56318). The Commission's related evaluation of these amendments is contained in a Safety Evaluation dated July 16, 2003.

No significant hazards consideration comments received: No.

Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County, Michigan

Date of application for amendment: February 13, 2003.

Brief description of amendment: The amendment revises Technical Specification (TS) 5.5.10, “Technical Specifications (TS) Bases Control Program,” to provide consistency with the changes to 10 CFR 50.59, which were published in the Federal Register (64 FR 53582) on October 4, 1999, and became effective March 13, 2001. Specifically, TS 5.5.10 has been revised to remove the phrase “unreviewed safety question.”

Date of issuance: July 22, 2003.

Effective date: As of the date of issuance and shall be implemented within 60 days.

Amendment No.: 156.

Facility Operating License No. NPF-43: The amendment revises the Technical Specifications.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28848). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 22, 2003.

No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: March 20, 2002, as supplemented by letter dated May 28, 2003.

Brief description of amendment: The amendment revises the reporting requirements specified in Section 2.E of the Facility Operating License and Technical Specification Section 5.6.4 by eliminating requirements that provide the U.S. Nuclear Regulatory Commission with information that is not risk significant, and change the reporting time period to be consistent with Section 50.73 of Title 10 of the Code of Federal Regulations.

Date of issuance: July 16, 2003.

Effective date: As of the date of issuance and shall be implemented 30 days from the date of issuance.

Amendment No.: 135.

Facility Operating License No. NPF-47: The amendment revised the Facility Operating License and Technical Specifications.

Date of initial notice in Federal Register: April 30, 2002 (67 FR 21286). The May 28, 2003, supplemental letter withdrew a portion of the original amendment request, but did not expand the scope of the original Federal Register notice or change the proposed no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 16, 2003.

No significant hazards consideration comments received: No.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: August 15, 2002, as supplemented by letter dated May 9, 2003.

Brief description of amendment: The amendment revises the reactor vessel surveillance program required by Title 10 of the Code of Federal Regulations, part 50, appendix H, section IIIB.3, allowing River Bend Station to incorporate the Boiling Water Reactor Vessel Internals Project Integrated Surveillance Program into the licensing basis.

Date of issuance: July 24, 2003.

Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.

Amendment No.: 136.

Facility Operating License No. NPF-47: The amendment consists of NRC staff approval of changes to the Updated Safety Analysis Report. Start Printed Page 46249

Date of initial notice in Federal Register: October 1, 2002 (67 FR 61679). The May 9, 2003, supplemental letter provided clarifying information that did not change the scope of the original Federal Register notice or the original no significant hazards consideration determination. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 24, 2003.

No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear Power Station, Plymouth County, Massachusetts

Date of application for amendment: July 5, 2002, as supplemented August 13, 2002.

Brief description of amendment: The amendment relocates portions of Technical Specification (TS) 3/4.6.B, “Primary System Boundary—Coolant Chemistry,” from the TSs to the Updated Final Safety Analysis Report (UFSAR). The portions of the TSs relocated to the UFSAR are the reactor coolant chemistry requirements for conductivity and chloride concentration. Specifically, TSs 3/4.6.B.2, 3/4.6.B.3, and 3.6.B.4 are relocated to the UFSAR.

Date of issuance: July 21, 2003.

Effective date: As of the date of issuance, and shall be implemented within 60 days.

Amendment No.: 202.

Facility Operating License No. DPR-35: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28850). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 21, 2003.

No significant hazards consideration comments received: No.

FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and 50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Beaver County, Pennsylvania

Date of application for amendments: October 31, 2001, as supplemented by letters dated December 21, 2001, and February 4, May 31, and December 2, 2002.

Brief description of amendments: The changes relocated the pressure temperature (P/T) limit curves and low temperature overpressure protection system limits to the Pressure and Temperature Limits Report (PTLR) in the BVPS-1 and 2 Licensing Requirements Manual and the reference that report in the affected TS limiting conditions for operation and Bases. The changes also included the addition of the PTLR to the Definitions Section of the TSs and added a new section to the reporting requirements in the Administrative Controls Section of the TSs delineating the necessary reports. The proposed changes were based on Generic Letter 96-03, “Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits,” dated January 31, 1996, and the Nuclear Regulatory Commission (NRC) staff's approval of the BVPS-1 and 2 plant-specific P/T limits methodology documented in the letter from Richard J. Laufer, NRC, to Mark B. Bezilla, FENOC, dated October 8, 2002.

Date of issuance: July 15, 2003.

Effective date: Effective as of the date of issuance and shall be implemented within 60 days.

Amendment Nos.: 256 and 138.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: December 26, 2001 (66 FR 66465). The supplements dated December 21, 2001, and February 4, May 31, and December 2, 2002, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 15, 2003.

No significant hazards consideration comments received: No.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: October 11, 2002, as supplemented March 4, 2003.

Brief description of amendment: The amendment revises Crystal River Unit 3 Improved Technical Specifications (ITS) 3.3.15, “Reactor Building Purge Isolation-High Radiation”; ITS Bases 3.7.15, “Spent Fuel Assembly Storage”; ITS 3.9.3, “Containment Penetrations”; and ITS 3.9.6, “Refueling Canal Water Level” to account for the handling of irradiated fuel within containment that has not occupied part of a critical reactor core within the previous 72 hours.

Date of issuance: July 14, 2003.

Effective date: As of the date of issuance, and shall be implemented prior to entering Mode 6 for the Cycle 13 refueling outage.

Amendment No.: 208.

Facility Operating License No. DPR-72: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 4, 2003 ( 68 FR 5676). The March 4, 2003, supplement contained clarifying information only and did not change the initial no significant hazards consideration determination or expand the scope of the initial application. The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 14, 2003.

No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit 1, Berrien County, Michigan

Date of application for amendment: December 10, 2002.

Brief description of amendment: The amendment consists of changes to the Donald C. Cook Nuclear Plant (D. C. Cook) Unit 1 Technical Specifications related to the reactor pressure vessel (RPV) operating limits at low temperatures. The amendment approves revised pressure-temperature limits for the RPV to be applicable for a maximum of 32 effective full-power years of facility operation. These changes were based, in part, on the use of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Case N-641.

Date of issuance: July 18, 2003.

Effective date: As of the date of issuance and shall be implemented prior to startup from Unit 1 refueling outage 19.

Amendment No.: 278.

Facility Operating License No. DPR-58: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: April 1, 2003 (68 FR 15762). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 18, 2003.

No significant hazards consideration comments received: No.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee Atomic Power Station, Lincoln County, Maine

Date of application for amendment: April 24, 2003.

Brief description of amendment: The amendment revises the Technical Specifications to eliminate the requirement for at least one person qualified to stand watch being present in the control room when irradiated fuel is stored in the fuel storage pool. Start Printed Page 46250

Date of issuance: July 02, 2003.

Effective date: Date of issuance to be implemented within [30] days from the date of issuance.

Amendment No.: 169.

Facility Operating License No. DPR-36: The amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28854). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 02, 2003.

No significant hazards consideration comments received: No.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County, Nebraska

Date of amendment request: October 8, 2002.

Brief description of amendment: The amendment changes the title of Shift Supervisor to Shift Manager. This amendment also replaces plant-specific titles with generic titles consistent with Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler TSTF-65, Rev. 1.

Date of issuance: July 15, 2003.

Effective date: As of the date of issuance and shall be implemented 60 days from the date of issuance.

Amendment No.: 200.

Facility Operating License No. DPR-46: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28854). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 15, 2003.

No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie Island Nuclear Generating Plant, Units 1 and 2, Goodhue County, Minnesota

Date of application for amendments: March 19, 2003.

Brief description of amendments: The amendments revise Technical Specification (TS) Section 5.3, “Plant Staff Qualifications.” The amendments update requirements that have been outdated based on licensed operator training programs being accredited by the National Academy for Nuclear Training and promulgation of the revised Title 10 of the Code of Federal Regulations, part 55, “Operators' Licenses.”

Date of issuance: July 22, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days.

Amendment Nos.: 159 and 150.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 15, 2003 (68 FR 18281). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 22, 2003.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: April 10, 2002.

Brief description of amendments: The license amendments revised several required actions in the Diablo Canyon Nuclear Power Plant (DCPP) Technical Specifications (TSs) that require suspension of operations involving positive reactivity additions or suspension of operations involving reactor coolant system (RCS) boron concentration reductions. In addition, the amendments revised several Limiting Condition for Operation notes that preclude reductions in RCS boron concentration when a reactor coolant pump(s) and/or a residual heat removal pump(s) are removed from operation. The changes allow small, controlled, safe insertions of positive reactivity, but limit the introduction of positive reactivity to ensure that compliance with the required shutdown margin or refueling boron concentration limits are satisfied.

Date of issuance: July 10, 2003.

Effective date: July 10, 2003, and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1-158; Unit 2-159.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: June 11, 2002 (67 FR 40024). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 10, 2003.

No significant hazards consideration comments received: No.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County, California

Date of application for amendments: February 28, 2003, as supplemented by letter dated June 26, 2003.

Brief description of amendments: The amendments revise Technical Specification (TS) Section 3.5.2, “ECCS—Operating,” Action A to allow a one-time increase in the allowed outage time for centrifugal charging pump (CCP) 1-1, for the purpose of seal replacement during Unit 1's Cycle 12 from 72 hours to 7 days. Additionally, the amendments delete a similar one-time TS change for Unit 2's CCP 2-1 that has expired.

Date of issuance: July 15, 2003.

Effective date: July 15, 2003, and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1—159; Unit 2—160.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: April 29, 2003 (68 FR 22753). The June 26, 2003, supplemental letter provided additional clarifying information that did not expand the scope of the application as originally noticed and did not change the staff's original proposed no significant hazards consideration determination.The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 15, 2003.

No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

Date of amendments request: February 11, 2003.

Brief Description of amendment: The amendment modifies Technical Specifications (TS) to allow a 40-month inspection interval for Farley, Unit 2 after the completion of the first post-replacement in-service inspection, rather than the completion of two consecutive inspections resulting in a classification of C-1.

Date of issuance: July 14, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment No.: 153.

Facility Operating License No. NPF-8: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: May 13, 2003 (68 FR 25657). The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 14, 2003.

No significant hazards consideration comments received: No.Start Printed Page 46251

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

Date of amendments request: March 21, 2003.

Brief description of amendments: The proposed Technical specifications (TS) amendments revise TS Section 5.5.1 “Offsite Dose Calculation Manual (ODCM).” The proposed change will remove reference to the Plant Operations Review Committee review and acceptance of licensee initiated changes to the ODCM.

Date of issuance: July 14, 2003.

Effective date: As of the date of issuance and shall be implemented within 30 days from the date of issuance.

Amendment Nos.: 160 & 152.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise the Technical Specifications.

Date of initial notice in Federal Register: May 27, 2003 (68 FR 28857).The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated July 14, 2003.

No significant hazards consideration comments received: No.

South Carolina Electric & Gas Company, South Carolina Public Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit No. 1, Fairfield County, South Carolina

Date of application for amendment: September 24, 2002, as supplemented April 8, 2003.

Brief description of amendment: This amendment revises the Technical Specification (TS) Surveillance Requirement 4.0.3, to incorporate the approved consolidated line item improvement program change associated with the TS Task Force traveler TSTF-358, “Change to Surveillance Requirement 3.0.3 Regarding Missed Surveillances.” Additionally, a change to the administrative controls section, Section 6.8, is included, to incorporate a new TS requirement for a Bases control program, consistent with the Bases control program presented in Section 5.5 NUREG 1431, “Improved Standard Technical Specifications for Westinghouse Plants,” Revision 2.

Date of issuance: July 11, 2003.

Effective date: As of the date of issuance and shall be implemented within 60 days from the date of issuance.

Amendment No.: 163.

Facility Operating License No. NPF-12: Amendment revises the Technical Specifications.

Date of initial notice in Federal Register: November 26, 2002 (67 FR 70768). The April 8, 2003, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the application.The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated July 11, 2003.

No significant hazards consideration comments received: No.

Start Signature

Dated at Rockville, Maryland, this 25th day of July 2003.

For The Nuclear Regulatory Commission.

Cornelius F. Holden,

Acting Director, Division of Licensing Project Management, Office of Nuclear Reactor Regulation.

End Signature End Preamble

[FR Doc. 03-19487 Filed 8-4-03; 8:45 am]

BILLING CODE 7590-01-P